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NEUTRON SPECTRUM DETERMIN TION BY

MULTIPLE FOIL CTIV TION METHOD

By

E

M.

ZSOLN Y

and

E.

J

SZONDI

Nuclear Training

Reactor

of

the Technical University Budapest

Received February 8 1982.

Presented by Dir. Dr. Gy. Csml

1 ntroduction

In case of determining neutron

flux

density spectra from multiple

foil

activation measurements one meets the problem of solving the following type

of activation integral equations:

x

A = SO ;(E)l/>(E)

d

i=

1(1)n

1)

o

where:

A

is

the experimentally determined saturation actIvIty per target

nucleus of the i

t

foil

detector extrapolated to infinite dilution

O ;(E) is

the energy dependent activation cross section

l/>(E)

is

the neutron flux density per unit energy interval

n -

is

the number of foil detectors used in the experiment.

In the practice one uses a discrete interval description of the neutron

energy range considered rather than a continuous representation by analytical

functions. Since the number of equations in this case

is

generally much smaller

than the number of unknowns that

is

the freedom of the system

is

rather high

the solution is not unique and the neutron flux spectrum cannot be derived

from these data without a priori assumption on the initial neutron spectrum

based on all physical informations available in the given case.

Several neutron spectrum unfolding computer codes [1-6 etc.J based

on

different philosophies have been used so far in the literature for the solution of

this problem. In the present report the program SANDBP developed in our

Institute

is

described [12J and the neutron spectrum in the centre of the core of

our nuclear reactor determined with aid of this program is presented.

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32

E

ZSOLSAY-E.

SZOSDI

2. The neutron spectrum

unfol ing

code SANDBP

This code

is

a version of the well-known program SAND Il, being

enlarged and modified in its algorithm. The use of this program involves

several advent ages compared to the old version:

a)

the activity and cross section errors can be taken into account in the

unfolding procedure;

b)

different response data deriving from the solution neutron spectrum

can be calculated

e.g.

neutron dose rate, neutron material damage

characteristics such as dpa, etc.);

c)

informations can be obtained

about

the magnitude of errors involved

in the solutions;

d)

the so called GOS-values characterizing the goodness of solution can

be determined;

e) the covariance

and

correlation matrices of the solution spectrum can

be determined;

f) the variety of output data

is

greater than that one in case of the old

program.

2 1 Principles of the

S ND

iterative method

The computer code S ND II has been developed to determine the

neutron

flux

density spectrum from activity measurements of foil detectors

activated in the neutron environment to be investigated [1]. The mathematical

procedure for the method involves the selection of an initial approximation

spectrum, and subsequent perturbation of that spectrum by iterative adjust

ments, to yield a solution spectrum producing calculated activities that agree

within a given error limit with the measured activities.

This way the code provides a best fit neutron flux density spectrum for a

given input set of infinitely dilute foil activities.

The iterative algorithm used in the program may be written as

j

=

l l)m

2)

where

L wt

In A ; / A ~ )

~ =

:...i=-=-l

)

j 1 1)m

3)

~ ·

.... I

i

and

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SELTROS

SPECTRL\f

DETER.lfISATJOS

cf } is the kth iterative neutron flux density assumed constant) over the

r

energy interval,

q

is the kth iterative flux density correction term for the r energy

interval,

A

-

is

the calculated saturation activity for the ith detector, based on the

kth

iterative neutron spectrum,

m - is the total number of energy intervals,

is

the number of detectors used.

The }j

is

a weighting function, which

is

in good approximation

proportional to the relative response for the

i

th

detector:

i=l l)n

j

l l)m

i

= l l)n

i=

1 1)/1

j

l l)m

4)

5)

6)

and

u i j

is the reaction cross section for the ith detector averaged constant) over

the energy interval.

The energy range of the solution spectrum is represented in max.

620

intervals depending on the cross section library. The standard library of the

program [lJ contains 620 energy groups from 10-

10

MeV to

18

MeV 45 per

decade up to 1 Me V

and 170

between 1 MeV and

18

Me

V).

The code includes

the provision for calculating flux density attenuation by foil cover materials

including cadmium, boron-10 and gold.

2 2 Characteristics of the program SANDBP

2.2.1. Weighting function and solution criteria

Due to the form of the weighting function defined by Eq.

4),

the structure

of the reponse function for a given

foil

detector therefore cross section

structure) may be reflected after several iterations in the solution spectrum

[1, 7, 8]. This effect may be pronounced especially that case if substantial

activity measurement errors or library cross section errors are present in the

data

set applied, and as a result

an

extended number of iterations

is

requested

to achieve an appropriate solution.

3

Periodica Polytechnica

El. 26/1-2

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34 t.

ZSOLSA}

--E_ SZOSDI

To

decrease

the

role

of

artificial

structures

in the solution spectrum

the

weighting function in Eq. 4) has been modified

taken

into account the activity

and

cross section errors, as follows:

Wk.= fj _1_

x

_ 1 _ )

l

A k

<5

A-)U (bu ..)

I \ J

i=1 1)n

j 1 1)m

7)

where (bAJ

and 6Ui)

are the relative

standard

errors

of

the activity

and

cross

section values, respectively. The exponents

u and u

are input parameters

and

can arbitrarily be chosen. This possibility enables the user to take into

consideration the quality of the input data

during

the unfolding procedure.

If u J 0

no

weighting with errors is used, that is the weighting function

agrees with

that

one

used in the

program

S i ~ ~ N

H.

As

the positive) value

of u and v

is increasing

the

solution spectrum will be

determined more and more

by

the more accurate

activity

and

cross section

data. Generally, if

the

errors

of the

different cross section values in the library

and/or

the activity

errors

within

the

detector set applied are

near

each other, a

value between 0.5

and

1 for

it

and

v is

reasonable. Otherwise a value between

1.5

and

2

is

recommended [10].

Too

large value:

of

these

parameters

equals

of

discarding the reactions having larger errors from the unfolding procedure the

relative weight

of

these reactions becomes

too

small, see Eqs 3)

and

7».

The

solution

criteria

of

the original

program

have also been extended.

The

standard

deviation of measured to calculated activity ratios is calculated after

each iteration step, expressed in

per

cents:

E

per ent

- - - - - ~ - - - - - - - - - - - - - - - - x 100

8)

and

in confidence interval units:

9)

where ~ i is the calculated

saturation

activity of the ith foil in the kth iteration

step) taken into account in the

latter

case the uncertainty of the experimental

data

The

iteration

is

then stopped

if

any

of

the

above

defined values is smaller

than

the one defined as input.

The

specification

of

larger

u

values in Eq. 7) will

increase the speed

of

the convergence defined by Eq.

9).

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SEUTROS

SPECTRU.lf

DETER \IINATIOS

35

The iteration procedure

is

ended also in

that

case ifthe standard deviation

becomes stable within less that

~ ~

in two successive iterations, or if the

maximum number of iterations specified by the input is achieved.

2.2.2. Activity run

The program involves the possibility to calculate detector activities for a

given neutron spectrum [1]. This procedure is performed in an ACTIVITY

run. The activities per target atom are calculated in units of Bq. consequently

they are the integrals of the reponse functions see point 2.1.).

2.2.3.

Calculation of response functions

Any response function reaction rate, dose rate, dpa-value, etc.) deriving

from the solution neutron spectrum can be calculated with aid of the program:

or

j l l)m

m

R=

L

P P

J

{E

j

1

-E

j 1

where R is a response function

Pj is the desired conversion factor.

10)

11)

Using this procedure neutron dose rate values can directly be calculated

from the neutron flux density spectrum eliminating the approximation errors

deriving from the calculation of average neutron energy over the spectrum

regions considered [9].

Nevertheless, the average neutron energy

E)

can also be determined,

using neutron density values as weighting functions rather than neutron flux

density values being involved in the original S ND II algorithm.

Thus

12)

where

j

is

the neutron flux value in the

/h

energy group.

3*

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36

E Z S O L ~ A Y E . S Z O ~ I

2.2.4. Monte-Carlo

error

analysis

a) Calculation of error limits

With aid of the SANDBP's Monte-Carlo program error analysis can be

performed. In the course of the Monte-Carlo runs the input activities

and

cross

sections (assumed that these values

fo11mv

a normal distribution) are varied

in accordance to their

random

errors:

i=1 1)n

where

~

-

is

the

/h

random

value of the

ith

measured activity Ai'

dAi -

is

the

standard

error of Ai'

9

is

a

random

number from normal distribution.

Similarly,

i = 1(1)/1

j= l l)m

(13)

(14)

where Ji

j

is

the rtn

random

value

of

the ith reaction cross section in the

j h energy group.

The cross section values are varied independently from each other in m

energy groups. l'Jo correlation is assumed between cross section values of

different energy groups:

i j l

= 0 for all

j and

I.

Performing the neutron spectrum unfolding procedure with these

modified input data, the relative standard error in ther energy group of the

solution spectrum can be written

as:

.; k 1 k

\

I ) Or,2 . . : : . -

(fy

\

J

I

:: .

. j / k . J)

-'

_ I r 1 r l

_

w._ X

k- l )x k -2) 1 k

k

L

C{Jj

r=1

j=1 1)/ll

(15)

where k is the number of Monte-Carlo runs. k 10 is allowed by the program.

The relative errors of the response data are derived on the analogy of the

Eq. (15).

As

it is seen from this equation the error limits calculated by the program

are simple

standard

errors (67.5 confidence level).

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S£[:TRON SPECTRG . .f DETERMISA

nos

.J i

b) Goodness of solution

In separate Monte-Carlo runs the input spectrum <P

inp

) is varied:

m

in p

. r

=

<Dip

p

x

-L

0 1

0 )

't - ) I.

e

j

l l)m

16

and the change of the output spectrum in the function of the input spectrum

is

used to characterize the goodness of solution (GOS):

j

= l l)m

(17)

Let the functionality between the point pairs (cp Jut and

~ o ~ n p

be approximated

by polynomial regression:

j = l(1)m 18)

,H-l

GOS

= )

(I

+

1)

x a

l

+ 1 X

<p}np)1

1=0

j=

l l)m

19)

where

Iv

is the degree of the poly nom. M 3 is used in the curve fitting

procedure.

GOS = 0 means

that

the

output

spectrum does not depend

on

the input

spectrum. A

GOS

value differing from zero indicates a worse solution.

The latter may happen in the energy regions poorly or not covered by the

response of the detector set applied. In this case the algorithm of the program

performs

an

interpolation,

and

the solution spectrum may simply reflect the

input spectrum.

c) Covariance and correlation matrices

As a result of the Monte-Carlo runs one has a series of solution spectra,

which can then be used to determine the covariance

and

correlation matrices

of

the nominal solution spectrum.

The value of covariance between the neutron flux density values of theph

and [th energy groups can be expressed as:

k

\ [ J ~ m r

L.,

, J t 1

r 1

S J I =

k

where

k

is the number of Monte-Carlo runs.

j =

l l)m

l=l l)m

20)

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38

1:

ZSOL;\ A

l - E SZOS I

The value of the correlation coefficient for the same relation:

t

~

j

1(l)m

1= l 1)m 21)

k

k

2.2.5. Code input

The program requires as input

a

the measured (input) activity values of the foil detectors corrected for

neutron selfshielding,

b cross section library for the detector set applied,

c

a guess neutron spectrum (input spectrum),

d relative errors of the activity and cross section values.

The SAND II program package contains a subroutine (VIF), which

allows the recording of user s problem data with a maximum freedom of

restrictive format. The input format and the input

data

deck sequence

description can be found in the user s guide [11].

2.2.6.

Code output

The program can supply the following outputs:

a In case of ITERATION (spectrum unfolding) run:

- listing of all cards of the input

data

set;

- if the solution

is

achieved a table summarizing the final iterative results

is given, including the tabulation of the solution neutron spectrum.

b In case of MONTE-CARLO run:

- all the above items are given followed by their relative errors;

- the GOS-values are written;

- covariance and correlation data of the solution spectrum are written

on a magnetic tape and they can also be printed.

c In case of ACTIVITY run items similar to those ones given in

ITERATION run can be obtained.

d utput to plotter: Due to corresponding specifications in the input

data set the solution neutron spectrum and GOS-values can be plotted.

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SEG TROS SPECTRU.\f DETERMIS TIOS

39

2.2.7.

Coding language and computer

The program

is

written in IBM

FORTR N IV (G)

language for an

1024

Kbyte R -40 type computer under the control of the operating system IBM OS

MVT. The code needs a memory region of

300

Kbyte.

3.

Neutron spectrum determination

The Budapest Technical University s nuclear reactor

is

a swimming pool

type light-water reactor with a maximum power of

100

kW [15]. The neutron

spectrum in the core position D5 (Fig.

1)

was determined.

For

this purpose sets

of activation detectors were irradiated

and

the neutron spectrum was unfolded

using the computer program SANDBP.

01

o

/

1 J l ~

/

A 8

(

0

D

o e

: ~ ~ : : ? : ;

o

c

o

Irradiation

Detector

Graphite

Fuel

Special graphite

with air gap)

Graphite with

a voi

F

G

H

n

--'i

o

tunn l

\

®

Manual control

ro

®

Automatic control ro

®

Safety ro

1001

Rabbit system

[ ]

rradiation channel

Water

Fig

1. Horizontal cross section of the

10

kW reactor core

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40

t

ZSOLA A Y-£ SZOSDJ

3 1 Experimemal procedure

The activation detectors were used in form of thin foils, irradiated in two

separate runs

at

a reactor power level of

10

kW

[9]. The foils were covered by

aluminium

or

cadmium, respectively. Three separate foils (Au, Co,

Ni)

with

different responses were used to monitor the reactor power leveL

The irradiated foils were counted by means

of

a Ge(Li) semi-conductor

detector and a 4096 channel pulse amplitude analyser system. The detection

peak efficiency curve of the spectrometer had previously been established.

The saturation activities of the radionuclides of interest, which serve as

the input of the spectrum unfolding program, are shown in Table

1.

Tabie 1

Saturation acth ities per target atom/or deteclOrs used

in

the Uti/aiding procedure

AS lvtarget

tom

Ll

Reaction Cover

Half-life ~ [ ; ; J

[Bq]

.1'31

1 6 ~ D y n , Y

65

Dy*

139.2 min 6.35 E-10 5.82

164Dy(n, i,),

65

Dy*

Cd

139.2 min 6.32 E-12

1.89

2 ~ A I n , p ) 2 ~ M g

Cd

9.46 min 9.01

E-16

12.49

1 Al(n, J . ) 2 ~ N a

Cd

15.03 h 1.60 E-16 4.88

45Sc(n.

i,)

6

S

C

*

Cd

84.0

d

3A5

E-12

0.63

56Fe(n.

p)56:Yln

Cd

2.587 h 1.35 E-17

5.78

58Ni(n. p)58mco

71 3 d 2A7 E-14 2.80

58Ni(n,

p)5smcO

Cd

71.3

d

2.48 E-14

7.17

115In(n,

i

6mIn

53 34

mm 1.75

E-il

1AO

115In(n. j.

m

In

Cd

53 34

min

L22 E-l1

0.82

55Mn(n.

;)

Mn

155.2 min

5.73 E-12

8.62

55Mn(n, {)56Mn

Cd

155.2 min 2.79 E-13

8.20

9

Au(n. ;V

9B

Au

2.695 d 5.86

E-li

2.94

197Au(n ,)198Au

Cd

2.695 d 3.22

E-l1

1.60

115In(n.

n')115mIn

Cd

4.5 h

4.56 E-14

1.30

47Ti(n,

p)

47

SC

Cd

80A

h

4.29 E-15

L13

4BTi(n,

p)'''BSC

Cd

44.1 h

3 92

E-17

0.74

Deleted from the unfolding procedure

3 2

Cross section library and neutron self hielding

The calculations were performed for 620 energy groups using the cross

section library

DOS ROSn

[13].

For

all the detectors the neutron

self-shielding correction was obtained by applying modified cross section

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SELTROS SPECTRu· \f DETERJflSATfOX

41

values in the cross section library, correcting each of the 620 group cross section

values of the reactions as follows:

22)

where:

23)

and r =

Nu}

is the thickness of the detector measured in units of mean free

path for absorption;

N

is

the number of target atoms per unit volume;

j

is the total cross section for goup

j;

t

is

the thickness of the detector;

E is

the exponential integral of the third order.

This approach enables us to introduce group cross sections corrected for

selfshielding. In absence of appropriate total cross section data capture cross

sections were used. To minimize the selfshielding effects thin dilutions of

detector materials were applied, except the threshold detectors.

3 3 nput neutron spectrum

The input neutron spectrum was a LWR type WWR reactor spectrum

available from reactor physics calculations in

49

energy groups [14J modified

in accordance with our former measuring results. Using the special inter-

polation

and

extrapolation procedure included in the unfolding program this

information was converted into a 620 groups spectrum. The joining point

between the Maxwellian and intermediate neutron energy region was

determined at energy E

'

2 x

10 7

MeV. The input spectrum

is

seen in Fig.

2.

3 4 Results

The reaction rates of

17

activation detectors were determined, however 4

reactions had to be deleted from the analysis showing incompatibility with

respect to the other detectors.

The aim of the investigations was, besides determining the actual form of

the neutron spectrum, to show the effect of the weighting function defined by

the program S NDBP on the unfolding procedure. For this purpose two

different runs were performed in conditions shown by Table 2.

In case of u= u= 0 no weighting with error values was used, while the

conditions u

= 2,

v

=

0 took the activity errors

of

the detector set into account

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42 E ZSOLNA

Y E.

SZONDI

05 Core

1016

.-----,.:----,------:..,-.----,---- --,....-----,---,---;----,---...,---,

Fig 2.

The input neutron spectrum

Table 2

lifolding conditions

and

integral parameters of the solution neutron spectra

No. of iterations

Ratio

of

measured to calculated

activities in confid. interval units

tPlolal

[m -1s 1 ]

tP>DleV

[m

2

s

1

]

tP .n leV [m 2S ]

tP>.2eV

[m

2

s

l

]

U=l =O

3

4.50

1.004E16

1.792E15

3.404E15

6.185E15

l I=2, r=O

3

3.79

1.008E16

1.817E15

3.562E15

6.596E15

during the unfolding procedure. As the difference between the error values of

the detector activities was rather large in the present case

see

Table 1.),

=

2

was chosen to ensure a bigger role for the more accurate reaction rates in

determination of the solution neutron spectrum [10]. The neutron spectra

obtained and some integral parameters referring to them are presented in

Figures 3 and 4 and in Table 2.

From the figures it is seen that the solution only slightly affects the

thermal part of the input spectrum but it changes the intermediate and fast

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0

:g

~

SEloTRON SPE TRUM

DETERMIS TlOS

5 ore

position without errors)

Differential spectrum obtained after 3 Iteration limit (

DPHI

/

DU

)

10

15

r - - - ~ - - - - ~ - - - - - - - - - - ~ - - - - ~ - - - - - . - - - - - . - - - - ~ - - - - ~ - - - - ~ - - - - __

~

i

~ 1 0 1 4

r - - - - - ~ - - f - - - - - - - ~ - - - - - L - - - - ~ - - - - _ . - - - - ~ - - - - - - - - - - - - ~ - - - - ~ - - - - ~

- T147[ R)SC47

~

INllS[NN)INllSM \

I 1

NIS8[N.P)C05SM, y ) ? O S8M

43

: i =::EJ.27[NP)MG27

10

13

L . : - - - - - - - L - - ~ - - - - - - . : - - - - - - - - - - _ _ _ ____ _: ____ ____ ______ O : A L " , ; 2 1 J . 7 : f [ N i , A ~ , ) t , t , N ~ A 1 . 2 q L : i ,

. .:'

10

10

10

9

10-

8

10

7

iQ-6 lO S 10-

4

10-

3

10-2 10-

1

100 101

Neutron energy [MeVl

Fig

3. Solution

neutron

spectrum

obt ined

after 3 iterations in case

of tI

=

u

=0

10-9

5

ore

position with act. errors)

Difierentiai spectrum obtained after

3

Iteration limit

DPHI/DU)

10 8 10

7

10

6

10-5 10-

4

10-3

Neutron energy

[MeV

J

I T147[ .P)SC47

INllS,[N.N)INllS_M \

NISS[Ntco58M

NISS[N,P)COS8M

_____.11

,

; i All 7(NP)MG

27

IAL

27(NA l NA24

1

10-2 10 1 100 10

1

Fig. 4. Solution

neutron spectrum obt ined

after 3

iter tions

in case of u=2

v o

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t

ZSOLNAY E.

SZONDI

energy region to a higher degree. Both solution spectra have some probably

largely artificial structure in the intermediate neutron energy region, but in

case of

t

2 the shape of the spectrum

is

smoother showing the effect of

weighting with the activity errors. This solution seems to be more realistic. In

the energy region (between

10

3 and about 1 ivleV poorly (or not) covered by

the response of the detector set applied the solution obtained with error

weighting ref1ects the input spectrum. However, in the other case some artificial

structure is introduced also in this part of the spectrum. In the fast energy

region, in the vicinity of 1 MeV, the effect of the oxygen resonance scattering

can be seen.

The intergral data ill Table 2 also sho\\ the difference between the two

solutions.

The correlation matrices for the solution neutron spectra are given in

Figures 5 and 6 in perspective representation. Strong correlation can be

observed between the flux density values

Fig 5. Correlation matrix for the solution neutron spectrum in case of

It

= =0

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SE TROS SPECTRUM DETERMJS TJOS

I l I l

therm l

r gion

Fig

6. Correlation matrix for the solution neutron spectrum in case of II 2. r 0

in the thermal

- in the intermediate

nd

- in the fast

neutron energy regions.

45

f

the activity errors are taken into account in the unfolding procedure

u 2, U

0

the solution

is

determined by the more accurate activity values. As

a result. the correlation matrix in this case has a less perturbed smoother

character compare Figures 5

nd

6 . The investigation

of

the structures in the

correlation matrices

nd

their interpretation is presently in progress.

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46

E ZSOL.VAY E.

SZONDI

Summary

The main features of the neutron spectrum unfolding code

SANDBP

are described

and

some results

obtained using this code are presented.

The program determines the neutron flux density spectrum from multiple foil activation

measurements using an iterative method. Response functions deriving from the unfolded neutron spectrum

can also be calculated. furthermore Monte-Carlo error analysis can be performed.

The neutron spectrum in the centre of the core of the Budapest Technical University s nuclear reactor

was measured and unfolded with aid of this program.

Acknowledgements

Thanks are given to

dr

Gy.

Csom

the director of the Institute for his interest

in

our work.

Thanks

are also due to Ms.

t. Sipos

for her assistance in the measurements

and

the computer work.

References

l.

McELROY

W

N. et

al : SAND

H.

Neutron

Flux Spectra Determinations by Multiple Foil Activation

Iterative Method. RSIC Computer Code Collection CCC-112. Oak Ridge National Laboratory.

Information Center, May

1969

2 FISCHER A : The

Petten Version of the

RFSP-JUL

Program.

R

M. G. Note 76/10. Petten, April 2nd,

1976

3 K A ~

F. B K et

al :

CRYSTAL BALL. A Computer Program for Determining Neutron Spectra from

Activation Measurements. ORNL-TM-4601. Oak Ridge National Laboratory, June

1974

4 ZUP

W. L et

al :

Comparison of Neutron Spectrum Unfolding Codes. The Influence of Statistical Weights

..in the SAND H Unfolding.

ECN

79-009, Petten,

January

1979

5 PEREY

F. G.: Least Squares Dosimetry Unfolding. The Program STAY SL. ORNL-TM-6062;

ENDF-

254.

6

Computer code WINDOWS,

ORNUTM-6656.

7

ZSOLNAY

E M.: Neutron Metrology in the L F.

R

Neutron Flux Density Spectrum in the Inner Graphite

Reflector of the

L

F.

R

ECN-77-134 Petten. October

1977

8 ZUP W

L

et

al :

Comparison of Neutron Spectrum Unfolding Codes. ECN-76-109, Petten. August 1976:

ECN-79-008. Petten. January 1979

9 ZSOLNAY

E M.-VIRAGH E.-SZONDl E J : Dosimetry Measurements in the Mixed (Neutron and

Gamma) Radiation Field of the Budapest Technical University s Nuclear Reactor. In: Proceedings of

the Third ASTM-Euratom Symposium on Reactor Dosimetry. Ispra (Varese) 1 to 5 October

1979

EUR 6813 EN-FR Vol. 1 (1980)

10

ZSOLNAY

E

M.-SzoNDlE

J :

Experiences with the

Neutron

Spectrum Unfolding Code SANDBP. BME

TR-RES-4/8l. Budapest. May

1981

11 SZONDl

E

1 :

User s Guide to the Neutron Spectrum Unfolding Code SANDBP.

To be

published.

[2.

SZONDl

E.

1.-ZsOl ,.\

E

M :

SA?\DBP. An Iterative Neutron Spectrum Unfolding Code. BME-TR

RES 2/81.

Budapest, March [981.

13

Zup W L.-NOLTHENIl S H. 1.-VAN

DER

BORG

NICO

J. C M : Cross Section Library DOSCROS77.

ECN-25. Petten, August [977.

14

Report

KFKI

76-76, Hungarian Academy of Sciences Central Research Institute for Physics, Budapest,

1976.

15 CSOM

Gy.: The Nuclear Training Reactor of the Budapest Technical University. Budapest, 1981.

Published in the Periodica Polytechnica El Eng. 26;1-2.

Eva

M ZSOLNAY

Dr. Egon J SZONDI

}

H-1521 Budapest

Documents

4744-8502-1-PB